WO2011026862A1 - Method for preparing a porous nuclear fuel containing at least one minor actinide - Google Patents

Abstract

The invention relates to a method for manufacturing a porous fuel including uranium, optionally plutonium, and at least one minor actinide, including the following consecutive steps: a) a step of compacting, into pellets, a mixture of powders including uranium oxide, optionally plutonium oxide, and at least one minor actinide oxide, at least a portion of the uranium oxide being in the form of triuranium octaoxide (U₃O₈), the other portion being in the form of uranium dioxide (UO₂); and b) a step of reducing at least one portion of the triuranium octaoxide (U₃O₈) into uranium dioxide (UO₂).

Description

 PROCESS FOR PREPARING A POROUS NUCLEAR FUEL BASED ON AT LEAST ONE MINOR ACTINIDE

DESCRIPTION

TECHNICAL AREA

 The invention relates to a method for preparing a porous nuclear fuel comprising uranium, possibly plutonium and at least one minor actinide implementing steps involving powders of these elements.

 This process can find, in particular, its application in the recycling of minor actinides via the incorporation of these minor actinides in the aforementioned fuel, which is intended to be used to constitute nuclear reactor nuclear rods or to enter the formation of transmutation targets, with a view to carrying out nuclear transmutation experiments, in particular to better understand the transmutation mechanism of these minor actinide elements.

For the rest of the talk, it is stated that minor actinide means actinide elements other than uranium, plutonium and thorium, formed in reactors by successive neutron captures by standard fuel nuclei, minor actinides being americium, curium and neptunium. STATE OF THE PRIOR ART

 At present, the recycling of minor actinides from the spent fuel processing is carried out by two distinct routes known as the following:

 - heterogeneous recycling; and

 - homogeneous recycling.

 In the case of heterogeneous recycling, the minor actinides are separated, during spent fuel, uranium and plutonium processing, and are then incorporated at a high level into fuel elements comprising a non-fissile matrix distinct from the Standard fuel elements of the reactor. Fuel elements comprising minor actinides may consist, for example, of cover elements disposed at the periphery of the core of a reactor. This recycling pathway makes it possible, in particular, to avoid degrading the characteristics of the standard fuel by incorporation of minor actinides by concentrating the problems generated by these actinides on a reduced flow of material.

In the case of homogeneous recycling, minor actinides are mixed at a low level and are distributed almost uniformly throughout the standard reactor fuel elements. To do this, during spent fuel treatment, uranium, plutonium and minor actinides are treated together to form oxides, which are then used in the manufacture of said fuels. The introduction of minor actinides into reactor core fuels results in significant emission of fission gases into these fuels and a high emission rate. For safety reasons, it is therefore necessary to provide fuels with a microstructure having a stable level of porosity under irradiation, which allows, moreover, the evacuation of fission gases and decay helium without physical degradation of the fuel . The recommended porosity rate for such fuels should be in the range of 14 to 16%, just as the porosity should be open porosity.

 To try to approach or even achieve such a rate, it is known to incorporate in the fuel significant amounts of organic pore-forming agents during the mixing / grinding step of the aforementioned oxides. However, the behavior over time of organic pore-forming agents is not ensured in view of the high level of emission generated by the presence of minor actinides. Indeed, the porogenic agents currently used (such as

Azodicarbonamide) very quickly lose their properties, which can generate a high scrap rate that is difficult to handle because of the presence of minor actinides. As a result, it is impossible to store the mixtures intended to constitute the fuels and, because of the degradation of the blowing agents, a risk of swelling of the pellets of the fuel before sintering intended to enter into the constitution of the fuel. It follows, therefore, an impossibility of obtaining fuels having a controlled porosity.

 Thus, in view of the processes of the prior art relating to the manufacture of fuel comprising minor actinides, the inventors have set themselves the objective of developing a process which does not have the drawbacks inherent in the use of organic porogenic agents, know the degradation of these agents from the stage of mixing fuel precursors and therefore an impossibility of obtaining a controlled porosity of the fuel.

STATEMENT OF THE INVENTION

 To solve this problem, the inventors propose an inventive method that makes it possible to dispense with the use of organic porogenic agents in order to obtain a porous nuclear fuel based on minor actinides.

 The invention thus relates to a method of manufacturing a porous fuel comprising uranium, possibly plutonium, and at least one minor actinide comprising successively the following steps:

a) a compacting step in the form of pellets of a mixture of powders comprising uranium oxide, optionally plutonium oxide, and at least one oxide of a minor actinide, at least a part of the the uranium oxide being in the form of triuranium octoxide U ₃ O ₈ , the other part being in the form of uranium dioxide UO ₂ ; b) a step of reducing at least a portion of the U ₃ O ₈ triuranium octoxide to uranium dioxide UO ₂ .

This process is innovative, among other things, by the fact that the porosity is generated by the reduction of triuranium octoxide U ₃ O ₈ to uranium dioxide UO ₂ · In fact, the reduction of triuranium octaoxide U ₃ O ₈ in uranium dioxide UO ₂ generates a reduction in volume of the space occupied by triuranium octaoxide by about 30%, the space left vacant constituting pores in the fuel. By regulating the amount of triuranium octoxide U ₃ O ₈ in the starting mixture and the reduction ratio, it is thus easily possible to aim for a precise rate of fuel porosity and thereby gain access to a method for controlling the porosity rate of the fuel produced.

 Moreover, because of the absence of use of organic pore-forming agents, the difficulties associated with the storage of the precursor mixtures of the fuel or the fuel pellets before sintering are less. The process of the invention is therefore more flexible than the processes of the prior art using organic pore-forming agents, since the various steps constituting it can be carried out extemporaneously.

Finally, this process makes it possible to envisage an incorporation of all the minor actinides (americium, curium and neptunium) resulting from spent fuel reprocessing streams. The oxide of a minor actinide may be americium oxide, such as AmO ₂ , ΑΠΙ 2Ο ₃ , curium oxide, such as Cm0 ₂ , Cm ₂ O ₃ , neptunium oxide, such as p0 ₂ and mixtures thereof.

 The plutonium oxide may be in the form of PuO 2 and / or PU 2 O 3.

In the aforementioned powder mixture, the uranium oxide in the form of triuranium octoxide U ₃ O ₈ can be associated within the same grains with an oxide of a minor actinide and optionally plutonium oxide, said grains advantageously having a grain size greater than 100 μm, preferably ranging from 100 to 250 μm. This does not exclude the fact that minor actinide oxide and plutonium oxide may also exist as separate grains.

 It is specified that, in what precedes and what follows, the term grain or powder having a grain size greater than 100 μm, a powder which, deposited on a sieve with square meshes of 100 μm side, will not cross said sieve.

 It is specified that, in what precedes and what follows, grain or powder having a grain size ranging from 100 to 250 μm, a powder that can be selected by the following sieving operations:

a first sieving operation on a sieve having square meshes of 100 μm in length making it possible to isolate the size fraction that does not pass through said sieve, that is to say the powder having a grain size greater than 100 μm; a second sieving operation of passing said powder having a grain size greater than 100 μm from the first sieving operation, on a sieve having square mesh of 250 μm on the side, the particle size fraction having passed through the sieve constituting the powder having a grain size ranging from 100 to 250 μm.

 Prior to compaction step a), the process of the invention may comprise a step of preparing the aforementioned powder mixture in step a).

 The preparation step of the mixture may, according to the invention, be carried out according to several variants.

According to a first variant, the step of preparing the powder mixture as defined in step a) may consist in bringing into contact a first mixture comprising a uranium oxide powder in the form of uranium dioxide UO ₂ , optionally a plutonium oxide powder, and at least one minor actinide oxide powder and a second powder mixture comprising uranium oxide in the form of U ₃ O ₈ triuranium octaoxide, optionally, plutonium oxide and an oxide of a minor actinide, the second mixture of powders being advantageously in the form of grains comprising the association within the same grain of the uranium oxide in the form of triuranium octoxide U ₃ O ₈ , optionally plutonium oxide, and an oxide of a minor actinide, said grains having advantageously a grain size greater than 100 μm, preferably ranging from 100 to 250 μm.

 The first mixture can be from the following operations:

an operation for contacting a uranium oxide powder in the form of uranium dioxide UO ₂ , optionally a plutonium oxide powder, and at least one powder of an oxide of uranium a minor actinide; and

 - optionally, a co-grinding operation of the resulting mixture to obtain a mixture of intimate powders.

 The second mixture can be from the following successive operations:

- an operation of contacting a uranium oxide powder in the form of Triuranium Octoxide U ₃ O _8, optionally a plutonium oxide powder, and at least one oxide powder minor actinide;

 an operation of co-grinding said powders;

 a pressing operation at a predetermined pressure P1;

 - a crushing operation; and

 at least one sieving operation so as to isolate the grains having a grain size greater than 100 μm, preferably ranging from 100 to 250 μm.

According to a second variant, the preparation step may consist in bringing into contact a first co-precipitated powder of uranium oxide, optionally of plutonium, and a minor actinide with a second mixture of powders comprising uranium oxide in the form of triuranium octoxide U ₃ O ₈ , optionally plutonium oxide and at least one oxide of a minor actinide.

 These powders may be derived from the oxalic coprecipitation of a stream comprising the chemical elements concerned.

 This second variant makes it possible to envisage using directly the reprocessing streams comprising the appropriate chemical elements in order to manufacture the fuel of the invention.

The second mixture of powders of the second variant (that is to say the one comprising, inter alia, uranium oxide in the form of triuranium octoxide U ₃ O ₈ ) can be obtained from a fraction of the first co-precipitated powder, which fraction is subjected to a step of calcination in air, so to transform the oxide of uranium UO ₂ triuranium octoxide U ₃ O _8, the resulting product being then optionally subjected to an operation pressing, followed by a crushing operation and a sieving operation so as to isolate the powders having a grain size greater than 100 μm, preferably from 100 to 250 μm.

From a concrete point of view, during the calcination operation, the formed UsOs dissociates to form distinct grains of triuranium octaoxide, the orthorombic form of the UsOs being incompatible with the cubic form of the oxide of plutonium and the oxide of a minor actinide present in the coprecipitated powder.

Whether for the first variant or for the second variant, the powders resulting from sieving operations (in this case the powders comprising triuranium octaoxide U ₃ O ₈ ), having a grain size of less than 100 μm ( that is to say, the powder which passes through a sieve having square meshes of 100 μm on one side) can be recovered and subjected to the following successive operations:

 a pressing operation at a pressure advantageously greater than 300 MPa;

 - a crushing operation;

 at least one sieving operation so as to isolate the powder having a grain size greater than 100 μm, preferably ranging from 100 to 250 μm,

 said powders being intended to enter into the constitution of the second mixture of powders.

 Whether for the first variant or for the second variant, the content of minor actinide in the powder mixture of step a) is advantageously in the range of up to 40% by weight relative to the total mass of the heavy nuclei (ie, U, Pu, minor Actinide (s)).

In the above-mentioned cases, the mixing steps can be carried out in an energy mixer, such as a mild mixer, for example of the Turbula type with oscillating-rotary movement or an oscillation mixer without grinding ball. The grinding steps can be performed in any type of mill such as, for example, a ball mill, attrition, oscillation, planetary motion or a gas jet mill.

 The sieving steps may be carried out by means of one or more sieves, for example a stainless steel sieve, making it possible to isolate the powder having a grain size greater than 100 μm, preferably a grain size ranging from 100 to 100 μm. 250 ym.

 The pressing steps can be performed by means of a press, for example hydraulic.

 The method of the invention then comprises a step of compacting the aforementioned mixture a) by pressing, to give the mixture a tablet shape, which will be the shape of the nuclear fuel pellets. This pressing step can be carried out at a pressure P2 equal to, lower than or greater than the above-mentioned pressure PI.

The resulting pellets are subjected to a reduction step in which all or part of the triuranium octoxide U ₃ O ₈ is reduced to uranium oxide UO ₂ , thereby concomitantly creating pores resulting from the vacant space left by the abovementioned reduction.

The reduction step may be carried out by subjecting the aforementioned pellets to a stream of reducing gas, for example hydrogen, optionally mixed with a neutral gas, such as argon at a temperature ranging from 600 to 1000 ° C for a duration ranging from 1 to 10 hours.

 Thus, it may be an argon and hydrogen mixture, the hydrogen being included in the mixture at a content of up to 5% by volume and optionally comprising water up to a content up to 20000 ppm.

This reduction step as already mentioned above generates a reduction of U ₃ O ₈ in UO ₂ and thus a reduction in volume. Advantageously, it will be possible to determine the amount of U ₃ O ₈ to be introduced so that after reduction, the subsequent porosity generates interconnected pores. To do this, it must be above the percolation threshold of the latter and take into account the volume reduction induced by the reduction of U ₃ O ₈ .

 The reduction step may be followed by a sintering step so as to consolidate between them the constituent grains of the pellets.

The sintering step may consist of heating the aforementioned pellets, for example, at a temperature ranging from 1000 to 1800 ° C, for a duration ranging from 1 to 8 hours, for example, under a neutral gas atmosphere, such as argon, optionally in the presence of hydrogen and water or in a reducing medium comprising hydrogen and optionally a neutral gas, such as argon, the hydrogen being included in the mixture to a content up to 5% by volume and optionally comprising water at a content of up to 20000 ppm. After the sintering step, the pellets obtained can be subjected to a grinding step, which can be carried out on a centerless and dry grinding machine, in order to obtain pellets satisfying the diameter specification.

 The fuel obtained by the process according to the invention has the following characteristics:

fuel having controlled porosity easily achievable by varying the amount of triuranium octoxide U ₃ O _{8 introduced} and its particle size;

 fuel whose porosity remains stable under irradiation;

 fuel which does not present a risk of degradation of organic pore-forming agents, as is the case with fuels comprising an organic pore-forming agent;

 fuel that can be stored over a long period.

 The invention will now be described with respect to the examples given below by way of illustration and not limitation.

BRIEF DESCRIPTION OF THE DRAWINGS

FIG. 1 is a thermal cycle diagram applied in the context of Comparative Example 1 and Example 1 of the invention showing the evolution of temperature T (in ° C) as a function of time t (in hours h ). FIG. 2 represents a view taken by optical microscopy (magnification * 2.5) for the material obtained according to comparative example 1.

 FIG. 3 represents a view taken by optical microscopy (magnification * 2.5) for the material obtained according to example 1 of the invention.

DETAILED PRESENTATION OF PARTICULAR EMBODIMENTS

COMPARATIVE EXAMPLE

 This example illustrates the preparation of a mixed oxide fuel comprising both uranium and americium.

 The method implemented according to this example comprises the following steps:

 a step of grinding a first mixture by co-milling 10 g of UO2 / AmO2 in a stainless steel bowl comprising two grinding balls in an oscillating mill for 30 minutes at 15 Hz;

 a step of adding 10 g of UO2 in the first mixture (respecting the ratio U / U + Am = 0.9) in a stainless steel bowl containing two grinding balls in an oscillating mill for 30 minutes at 15 Hz ;

a storage step of the UO2 / AmO2 mixture in a stainless steel container, this mixture resulting from the addition of the 10 g of UO ₂ ;

 a step of pressing the resulting mixture UO 2 / AmO 2 at a pressure P 1 of 400 MPa by means of a tri-shell matrix with lubrication of the matrix and the pistons (with stearic acid);

a reduction step under Ar / ¾ at 1000 ° C for 1 hour; a natural sintering step under 4% hydrogenated argon at 1750 ° C. for 4 hours.

 The reduction / sintering thermal cycle is represented, more explicitly, in FIG. 1 representing the evolution of the temperature as a function of time (in hours).

At the end of this example, a pellet of formula U0.9 m ₀ , ι0 ₂ - _χ having a porosity of approximately 5% (it is a geometric density) and having a open porosity rate much less than 5% determined by hydrostatic weighing.

 The pellet obtained was analyzed by light microscopy and X-ray diffraction.

 A snapshot obtained by light microscopy is shown in FIG.

 It is observed in this picture that the material has a very low level of porosity.

 The geometric density of the material is estimated at a value of 95% (geometric density obtained by weighing and by means of a laser profilometer).

 The mass loss of the material after heat treatment is estimated at 3.5%.

EXAMPLE 1 OF THE INVENTION

 This example illustrates the preparation of a mixed oxide fuel comprising both uranium and americium according to the process of the invention.

 The method implemented according to this example comprises the following steps:

a step of preparation of a first mixture by co-grinding of 10 g U0 ₂ / m ₂ (according to a ratio U / U + Am = 0.9) in a stainless steel bowl comprising two grinding balls in an oscillating mill for 30 minutes at 15 Hz;

a step of preparing a second mixture U3O8 / A1I1O2 with a mass of 10 g while respecting the ratio U / U + Am = 0.9 in a stainless steel bowl containing two grinding balls in an oscillating mill for 30 minutes to 15 Hz. U ₃ O ₈ is introduced at 40% by weight relative to the mass of the final mixture;

 a storage step of the first mixture

U02 / Am02 in stainless steel container;

 in parallel, a step of pressing the second mixture U3O8 / A1I102 at a pressure P1 of 400 MPa by means of a tri-shell matrix with lubrication of the matrix and the pistons (with stearic acid);

 a step of crushing the pellets resulting from the pressing of the mixture U3O8 / A1I1O2;

a step of sieving the U3O8 / Al2O2 mixture, so as to isolate the particle size fraction having a grain size ranging from 100 to 250 μm, said constituent grains of said fraction comprising within each grain the U ₃ 0 ₈ / AmO combination; ₂ ;

a mixing step (without grinding means in an oscillating mill at 15 Hz for 30 minutes) of the first mixture U0 ₂ / Am0 ₂ and a sample of the second mixture U3O8 / A1I102 thus sieved, so as to guarantee 40% in mass of U ₃ O ₈ in the final mixture;

a step of pressing the resulting mixture at a pressure P2 of 400 MPa; a reduction step under Ar / ¾ at 1000 ° C for 1 hour so as to reduce U3O8 to UO2;

 a natural sintering step under 4% hydrogenated argon at 1750 ° C. for 4 hours.

 The reduction / sintering thermal cycle is represented, more explicitly, in FIG. 1 representing the evolution of the temperature as a function of time (in hours).

At the end of this example, a lozenge of formula Uo 9 m ₀ , 10 2 having a porosity of approximately 14% (it is a geometric density) and having 10% of porosity is obtained. open determined by hydrostatic weighing.

 The pellet obtained was analyzed by light microscopy and X-ray diffraction.

 A photograph obtained by light microscopy is shown in FIG.

 It is observed in this picture that the porosity of the material obtained has an elongated morphology of the lenticular type and that it is mostly interconnected.

 The geometric density of the material is estimated at a value of 86% (geometric density obtained by weighing and by means of a laser profilometer).

 The mass loss of the material after the heat treatment is estimated at 5%.

By X-ray analysis, a disappearance of the U ₃ O ₈ phase can be observed after heat treatment (which represents the complete reduction of U ₃ O ₈ in UO ₂ ) · In addition, the O / Am ratio of the oxide Americium is in the range 1.5 <0 / Am <2 because no peak of Am ₂ C> 3 or AmO ₂ was observed. This also reflects the fact that the material is a single-phase material (which means that americium and uranium are mixed at the atomic scale).

Claims

A method of manufacturing a porous fuel comprising uranium, optionally plutonium, and at least one minor actinide comprising successively the following steps: a) a compacting step in the form of pellets of a mixture of powders comprising uranium oxide, optionally plutonium oxide, and at least one oxide of a minor actinide, at least a part of the the uranium oxide being in the form of triuranium octoxide U ₃ O ₈ , the other part being in the form of uranium dioxide UO ₂ ; b) a step of reducing at least a portion of the U ₃ O ₈ triuranium octoxide to uranium dioxide UO ₂ . The manufacturing method according to claim 1, wherein the oxide of a minor actinide is selected from an americium oxide, a curium oxide, a neptunium oxide and mixtures thereof. 3. Manufacturing method according to any one of the preceding claims, wherein the uranium oxide in the form of triuranium octaoxide U ₃ O ₈ is associated within the same grains with an oxide of a minor actinide and optionally plutonium oxide, said grains having a grain size greater than 100 μm, preferably from 100 to 250 μm. 4. Method according to any one of the preceding claims, further comprising, prior to compaction step a), a step of preparing said powder mixture as defined in step a). 5. Method according to claim 4, wherein the step of preparing the powder mixture comprises contacting a first mixture comprising a uranium oxide powder in the form of uranium dioxide UO ₂ , optionally a powder of plutonium oxide, and at least one powder of an oxide of an actinide minor and a second powder blend comprising uranium oxide in the form of triuranium octoxide U ₃ O _8, optionally oxide of plutonium and an oxide of a minor actinide. 6. The method of claim 5, wherein the second powder mixture is in the form of grains comprising the combination within the same grain of the uranium oxide in the form of triuranium octaoxide U ₃ O ₈ optionally, plutonium oxide, and an oxide of a minor actinide, said grains having a grain size greater than 100 μm, preferably from 100 to 250 μm. 7. The method of claim 5, wherein the first mixture is derived from the following operations: an operation for contacting a uranium oxide powder in the form of uranium dioxide UO ₂ , optionally an oxide powder of plutonium, and at least one powder of a minor actinide oxide; and  optionally a co-grinding operation of the resulting mixture to obtain an intimate mixture of powders. 8. The method of claim 6, wherein the second powder mixture is derived from the following operations: - an operation of contacting a uranium oxide powder in the form of Triuranium Octoxide U ₃ O _8, optionally a plutonium oxide powder, and at least one oxide powder minor actinide;  a co-grinding operation of said powders;  a pressing operation at a predetermined pressure P1;  - a crushing operation; and  at least one sieving operation so as to isolate the grains having a grain size greater than 100 μm, preferably ranging from 100 to 250 μm. 9. The method of claim 4, wherein the step of preparing the powder mixture comprises contacting a first co-precipitated powder of uranium oxide, optionally plutonium, and a minor actinide with a second mixture of powders comprising uranium oxide in the form of triuranium octoxide U ₃ O ₈ , optionally plutonium oxide and at least one oxide of a minor actinide. 10. The method of claim 9, wherein the second powder mixture is obtained from a fraction of the first co-precipitated powder, which fraction is subjected to a calcination step in air, so as to transform the oxide of uranium UO ₂ in triuranium octoxide U ₃ O ₈ , the resulting product then optionally being subjected to a pressing operation, followed by a crushing operation and a sieving operation so as to isolate the powders having a grain size greater than 100 μm, preferably from 100 to 250 μm. 11. The method of claim 8 or 10, wherein the powders from sieving operations having a grain size of less than 100 μm are recovered and subjected to the following successive operations:  a pressing operation at a pressure advantageously greater than 300 MPa;  - a crushing operation;  at least one sieving operation so as to isolate the powder having a grain size greater than 100 μm, preferably ranging from 100 to 250 μm, said powders being intended to enter into the constitution of the second mixture of powders. 12. The method as claimed in claim 1, in which the reduction step b) is carried out by passing a stream of reducing gas at a temperature ranging from 600 to 1000 ° C. for a duration that may range from 1 to 10 hours. 13. A method according to any one of the preceding claims, further comprising, after the step of reducing b), a step of sintering the fuel pellets. 14. The method of claim 13, wherein the sintering step is carried out by heating the aforementioned pellets at a temperature ranging from 1000 to 1800 ° C, for a period ranging from 1 to 8 hours. 15. The method of claim 14, wherein the sintering step is carried out in a neutral gas atmosphere, optionally in the presence of hydrogen and water.